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DOE OSTI2011

OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375 kg PWR core melt, initially containing 15 wt% siliceous concrete, with a specially designed two-dimensional siliceous concrete test section with an initial cross-sectional area of 50 cm x 50 cm. The sand and aggregate constituents for this particular siliceous concrete were provided by CEA as an in-kind contribution to the program. The report begins by providing a summary description of the CCI-3 test apparatus and operating procedures, followed by presentation of the thermal-hydraulic results. Detailed posttest debris examination results will be provided in a subsequent publication. Observations drawn within this report regarding the overall cavity erosion behavior may be subject to revision once the posttest examinations are completed, since these examinations will fully reveal the final cavity shape.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

OECD MCCI project final report, February 28, 2006.

Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. The fractured crust will provide a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed and contribute to terminating the core-concrete interaction. Thus, one of the key aims of the current program was to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit, the existing reactor material database for dry cavity conditions is solely one-dimensional. Although the MACE Scoping Test was carried out with a two-dimensional concrete cavity, the interaction was flooded soon after ablation was initiated to investigate debris coolability. Moreover, due to the scoping nature of this test, the apparatus was minimally instrumented and therefore the results are of limited value from the code validation viewpoint. Aside from the MACE program, the COTELS test series also investigated 2-D CCI under flooded cavity conditions. However, the input power density for these tests was quite high relative to the prototypic case. Finally, the BETA test series provided valuable data on 2-D core concrete interaction under dry cavity conditions, but these tests focused on investigating the interaction of the metallic (steel) phase with concrete. Due to these limitations, there is significant uncertainty in the partitioning of energy dissipated for the ablation of concrete in the lateral and axial directions under dry cavity conditions for the case of a core oxide melt. Accurate knowledge of this 'power split' is important in the evaluation of the consequences of an ex-vessel severe accident; e.g., lateral erosion can undermine containment structures, while axial erosion can penetrate the basemat, leading to ground contamination and/or possible containment bypass. As a result of this uncertainty, there are still substantial differences among computer codes in the prediction of 2-D cavity erosion behavior under both wet and dry cavity conditions. Thus, a second key aim of the current program was to provide the necessary data to help resolve these modeling differences. In light of the above issues, the OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program was initiated at Argonne National Laboratory. The project conducted reactor materials experiments and associated analysis to achieve the following technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focused on providing both confirmatory evidence and test data for the coolability mechanisms identified in previous integral effects tests, and (2) address remaining uncertainties related to long-term 2-D core-concrete interaction under both wet and dry cavity conditions. Data from the various tests conducted as part of the program is used to develop and validate models and codes that eventually form the basis for extrapolating the experimental findings to plant conditions. Achievement of these technical objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs of future plants. The project completed a total of eleven reactor material tests to investigate melt coolability and 2-D core-concrete interaction mechanisms under both wet and dry cavity conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Proposed Activities to Address Regulatory Gaps and Challenges for Licensing Advanced Reactors Using Seismic Isolation

Over the last decade, particularly since implementation of the certified design regulatory approaches outlined in 10 CFR 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” interest has been increasing in the use of seismic isolation (SI) technology to support seismic safety in nuclear facilities. In 2009, the United States (U.S.) Nuclear Regulatory Commission (NRC) initiated research activities to develop new guidance targeted at isolated facilities because SI is being considered for nuclear power plants in the U.S. One product of that research, which was developed around a risk-informed regulatory approach, is a draft NRC NUREG series (NUREG/CR) report that investigates and discusses considerations for use of SI in otherwise traditionally founded large light water reactors (LWRs). A coordinated effort led to new provisions for SI of LWRs in the American Society of Civil Engineers standard ASCE/SEI 4-16, “Seismic Analysis of Safety Related Nuclear Structures.” The risk-informed design philosophy that underpinned development of the technical basis for these documents led to a set of proposed performance objectives and acceptance criteria intended to serve as the foundation for future NRC guidance on the use of SI and related technology. Although the guidance provided in the draft SI NUREG/CR report and ASCE/SEI 4 16 provides a sound basis for further development of nuclear power plant designs incorporating SI, these initial documents were focused on surface-founded or near-surface-founded LWRs and were, necessarily, limited in scope. For example, there is limited information in both the draft NUREG/CR report and ASCE/SEI 4-16 related to nonlinear analysis of soil-structure systems for deeply-embedded reactors, the isolation of components, and the use of vertical isolation systems. Also not included in the draft SI NUREG/CR report are special considerations for licensing of isolated facilities using the certified design approach in 10 CFR 52 and a detailed discussion of seismic probabilistic risk assessments for isolated facilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of an ejecta particle size measurement diagnostic based on Mie scattering

The goal of this work is to determine the feasibility of extracting the size of particles ejected from shocked metal surfaces (ejecta) from the angular distribution of light scattered by a cloud of such particles. The basis of the technique is the Mie theory of scattering, and implicit in this approach are the assumptions that the scattering particles are spherical and that single scattering conditions prevail. The meaning of this latter assumption, as far as experimental conditions are concerned, will become clear later. The solution to Maxwell’s equations for spherical particles illuminated by a plane electromagnetic wave was derived by Gustav Mie more than 100 years ago, but several modern treatises discuss this solution in great detail. The solution is a complicated series expansion of the scattered electric field, as well as the field within the particle, from which the total scattering and absorption cross sections as well as the angular distribution of scattered intensity can be calculated numerically. The detailed nature of the scattering is determined by the complex index of refraction of the particle material as well as the particle size parameter, x, which is the product of the wavenumber of the incident light and the particle radius, i.e. x = 2rπ= λ. Figure 1 shows the angular distribution of scattered light for different particle size parameters and two orthogonal incident light polarizations as calculated using the Mie solution. It is obvious that the scattering pattern is strongly dependent on the particle size parameter, becoming more forward-directed and less polarizationdependent as the particle size parameter increases. This trend forms the basis for the diagnostic design.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Evaluating Corrosion Effects on the Stainless Steel Components of the SAVY-4000/Hagan Nuclear Material Storage Containers

Surveillance of vented, stainless-steel Hagan and SAVY-4000 storage containers with some plutonium materials has revealed evidence of corrosion inside the container. There are ~3000 items stored in Hagan containers and ~600 items currently stored in SAVY-4000 containers at TA -55. The Hagan is made of 304L steel, while the SAVY-4000 is made of 316L and should be less susceptible to corrosion. Given the relatively thin body wall of these containers (nominally 0.762 millimeters), it is essential to understand the effects of corrosion on the lifetime of the containers to ensure worker safety. Studies are being conducted on both container types to determine the relative susceptibility of the two containers to corrosion. This paper describes the potential cause of corrosion, the extent of corrosion, and types of corrosion observed, along with the results of accelerated aging experiments. The combined results of the surveillance and lifetime extension studies establish a technical basis for extending the design life of the stainless steel components of the SAVY-4000 container from 5 years to 15 years.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Cost and Performance Baseline for Fossil Energy Plants - Volume 4: Coal-to-Liquids via Fischer-Tropsch Synthesis

This report establishes performance and cost data for coal-to-liquids systems, specifically by means of gasification and Fischer-Tropsch reaction. The analyses were performed on a consistent technical and economic basis to assess the design and financial performance of a commercial-scale coal-to-Fischer-Tropsch liquids facility. The cost and performance data were compiled from published reports, information obtained from vendor quotes and users of the technology, and data from designing and building utility and refining projects.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2018

Continuation Report: Creep-fatigue Behavior and Damage Accumulation of a Candidate Structural Material for Concentrating Solar Thermal Receiver

At the start of this phase, six alloys were determined to be candidate materials for the concentrating solar power (CSP) thermal receiver: Alloys 617, 625, 740H, 230, 282 and Alloy X. From these, Alloys 740H and 282 were determined to be the most suitable, with both alloys exhibiting similar high temperature strength properties. Alloy 740H was selected due to its ease to work with (simpler heat treatments and welding procedures) and a larger database of existing data. Fatigue testing was started in this phase, with four fatigue tests having been completed, with additional fatigue testing planned, as well as tensile, creep and creep-fatigue tests starting in the next phase. Two fatigue tests performed at 850 °C and were completed in significantly fewer cycles than the two tests performed at 750 °C. Metallography was performed on the fatigued specimens to examine crack path and surrounding microstructure. These preliminary results suggest a significant drop in strength at elevated temperatures above 750 °C, however, more test results will be needed prior to setting design limits. Several potential model high temperature design methods were evaluated that could form the basis of our final recommended creep-fatigue design method for Alloy 740H. To do this we expanded on the receiver finite element model to develop a model representative of a potential Gen 3 design using KCl-MgCl2 salt coolant and operating with a salt inlet temperature of 550°C and salt outlet temperature of 720°C. Using the temperature fields from the model thermohydraulic analysis we determined the design life of the reference receiver assuming it was constructed from 740H seamless tubing and assuming high temperature design properties for 740H from literature data and, where required, naïve extrapolation. As experimental data is generated, these tentative design parameters will be updated to match the actual performance of 740H in elevated temperature service. The results of this design study are a rank ordering of potential design methods for two criteria: ease of use and design margin. The end goal of this project is to develop a creep-fatigue design method with adequate, but not over conservative, design margin that is straightforward to execute. Our receiver model is based on the preliminary design provided to DOE by Solar Reserve. Our analysis indicates a peak metal temperature of around 840° C for this design (comparable to Solar Reserve’s analysis, which shows a peak metal temperature of around 830° C). The Solar Reserve tube thickness is too thin to sustain creep-fatigue loading at this temperature. This means that the current trial design is not well-suited for evaluating different design methodologies, as we predict it will have a very short design life. We suggest that we modify our reference design with input from DOE in order to lower the maximum metal temperature to less than 800° C. We can then repeat these analyses and give a more meaningful comparison of different design methods before making a final selection.

14 SOLAR ENERGY↗
DOE OSTI2020

Qubit Based Single-Photon Sensors for Dark Matter Searches

This experimental program will adapt low-noise qubit readout techniques for use in particle physics experiments. As an end goal, this research will deliver a technical demonstration of low-noise quantum sensors, as well as the results of a proof-of-principle axion search in a narrow but previously unexamined mass range. This program would form the basis for a technical design report outlining a next-generation axion dark matter search.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2020

Guide for Joint Participation in a DOE Complex Wide SAVY-4000 Surveillance Program

This document describes LANL Storage Program best practices when working with other User Sites to implement DOE M441.1-1 Manual requirements through the use of the SAVY nuclear material packaging system. It is a Manual directive for Field Element Managers to review and approve the SAVY-4000 container series design (with supporting technical basis) as a packaging system for use at their respective sites. This review and approval incorporates an evaluation of the User Site methodology for SAVY surveillance and the process for documenting its nuclear materials storage program. This document aims to assist the Field Element Mangers and User Sites in the processes necessary to credit LANL SAVY surveillance for key elements of the User Site surveillance plan.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

HFIRCON Version 1.0.5 User Guide

The High Flux Isotope Reactor (HFIR) Controller (HFIRCON) code is a collection of python routines and C plugins that automate the workflow for fuel and single- or multicycle target depletion analyses for HFIR at Oak Ridge National Laboratory (ORNL). This code calls the LAVAMINT (LAVA Model Interrogator) parallel (MCNP) Monte Carlo N-Particle model interrogator to stochastically calculate cell volumes and bounding boxes, the ADVANTG (Automatic Variance Reduction Generation) code package for all variance reduction and source biasing calculations, the ORNL-Transformative Neutronics/MCNP5 transport solver for all transport solutions, and the MSX_DEPLETE module to perform all depletion calculations via the ORIGEN (Oak Ridge Isotope Generation) application programming interface. It also performs a robust set of postprocessing functions to automatically provide summaries of several key metrics that are common to a wide variety of typical HFIR design and safety-basis analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Extension of SCALE/Sampler’s sensitivity index for the assessment of cross section, fission yield, and decay data uncertainties

Accurate prediction of nuclide compositions and neutron and gamma sources and spectra for fresh and spent nuclear fuel through computational modeling and simulations is the basis for safeguards instruments design, optimization, and calibration, and for validation of the measured responses. The accuracy of these simulations can be significantly impacted by uncertainties in input parameters to the applied computer code. Input parameters are, for example, dimensions, material compositions, and temperatures of the model. Additionally, important but sometimes neglected input parameters for simulations of nuclear systems are the nuclear data that include nuclear reaction cross sections, fission product yields, and decay data. It has been shown that uncertainties of calculated results are dominated by uncertainties of these nuclear data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The Architectural Dynamics of the Bacterial Flagellar Motor Switch

The rotary bacterial flagellar motor is remarkable in biochemistry for its highly synchronized operation and amplification during switching of rotation sense. The motor is part of the flagellar basal body, a complex multi-protein assembly. Sensory and energy transduction depends on a core of six proteins that are adapted in different species to adjust torque and produce diverse switches. Motor response to chemotactic and environmental stimuli is driven by interactions of the core with small signal proteins. The initial protein interactions are propagated across a multi-subunit cytoplasmic ring to switch torque. Torque reversal triggers structural transitions in the flagellar filament to change motile behavior. Subtle variations in the core components invert or block switch operation. The mechanics of the flagellar switch have been studied with multiple approaches, from protein dynamics to single molecule and cell biophysics. The architecture, driven by recent advances in electron cryo-microscopy, is available for several species. Computational methods have correlated structure with genetic and biochemical databases. The design principles underlying the basis of switch ultra-sensitivity and its dependence on motor torque remain elusive, but tantalizing clues have emerged. This review aims to consolidate recent knowledge into a unified platform that can inspire new research strategies.

59 BASIC BIOLOGICAL SCIENCES↗
NASA NTRS1971

Generalized safety equation - A concept

Concept provides definition of relationship between safety and reliability, personnel safety measurement, and equipment safety evaluation. Safety systems cope with single or combined risks. Cost and effectiveness of alternate hypothetical safety systems are estimated and used as basis for final system design.

Hano, G.↗
NASA NTRS1972

Space vehicle gyroscope sensor applications

Considerations which form the basis for the specification, design and evaluation of gyroscopes for spaceborne sensor applications are presented. The applications are distinguished by basic vehicle category: launch vehicles, spacecraft, entry vehicles and sounding rockets. Specifically excluded from discussion are gyroscope effector applications. Exotic or unconventional gyroscopes for which operational experience is nonexistent are mentioned only briefly to alert the reader of future trends. General requirements for testing and evaluation are discussed.

Source record↗
NASA NTRS1972

Nacelle design studies for advanced transport aircraft.

Results are given of several analytical studies of nacelles suitable for advanced subsonic commercial transport aircraft. The impact on the nacelle of reduced aircraft noise and increased cruise Mach number is emphasized and initially developed in terms of the individual nacelle components: inlet, fan cowl, nozzle, etc. This is achieved by relating the noise and cruise speed constraints to which the aircraft system must be designed to specific limitations on the individual nacelle components. Performance assessments are then made (separately for each nacelle component) of competitive design concepts. Overall nacelle designs, synthesized on the basis of the individual component studies, are briefly discussed.

Sussman, M. B.↗
NASA NTRS1974

Booster aerodynamic heating: Test support

Several technical areas were encompassed in providing support for booster thermal environment test work. These areas included: (1) cavity flow heating, (2) rarefied flow heating, and (3) impulse operated model research and testing. Cavity flow heating problems were studied with respect to the proposed altitude control motors for the space shuttle. Available literature on this subject was reviewed and analytical predictive methods were summarized for use in planning testing work. Rarefied flow heating data was reviewed and correlated. The study showed the importance of considering rarefied flow conditions in launch thermal environment prediction. Impulse operated model research and testing was conducted to provide a basis for understanding and designing such models for booster thermal environment testing.

Engel, C. D.↗
NASA NTRS1974

Vacuum distillation/vapor filtration water recovery

The development and evaluation of a vacuum distillation/vapor filtration (VD/VF) water recovery system are considered. As a functional model, the system converts urine and condensates waste water from six men to potable water on a steady-state basis. The system is designed for 180-day operating durations and for function on the ground, on zero-g aircraft, and in orbit. Preparatory tasks are summarized for conducting low gravity tests of a vacuum distillation/vapor filtration system for recovering water from urine.

Honegger, R. J.↗